An approach has been developed to model in a simple way count loss in Passive Neutron Coincidence and Multiplicity Counting (PNCMC) systems in order to determine dead time corrections. The approach ...does not require to simulate the full PNCMC system, but rather uses basic information from the PNCMC system such as the neutron detection efficiency, the counters cabling scheme and the dead times of different electronic components of the system. A good agreement is found between the measured dead time parameters of a neutron multiplicity counter described in the literature and the dead time parameters calculated using the presented approach.
Program title: COLONEMA
Catalogue identifier: AEYS_v1_0
Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEYS_v1_0.html
Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland
Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html
No. of lines in distributed program, including test data, etc.: 2900
No. of bytes in distributed program, including test data, etc.: 18753
Distribution format: tar.gz
Programming language: C++.
Computer: Unix/Linux workstations and PC.
Operating system: Unix, Linux and windows, provided a C++ compiler has been installed. Examples were tested under Debian Linux.
RAM: Depends on the simulated neutron source strength and detector efficiency. !at most, the example presented in the article allocates about 2 GB of virtual memory.
Classification: 11.7, 17.7.
Nature of problem: To determine the mass of fissile materials inside an unknown object, a passive neutron multiplicity or coincidence counting is performed in which single, double and triple neutron rates are measured. The measured rates must however be corrected from count losses due to dead time when the count rate is large.
Solution method: Unlike approaches based on time consuming Monte Carlo transportation of the neutrons in the inspection system combined with treatment that take into account count losses due to dead time, a versatile Monte Carlo approach is presented which uses only the detection efficiency of the system and the dead time of the electronics and counters.
Running time: It takes about 109 min on an Intel Xeon X5550 2.676 GHz processor to run the example presented.
Photofission provides a mean for measuring nuclear material in large and dense radioactive waste packages. High-energy interrogating bremsstrahlung photons produced by a LINAC are indeed able to ...induce fissions in the depth of the waste matrix. In addition, a scintillation detector can detect high-energy delayed gamma rays that are able to escape the waste package. Scintillation detectors also allow performing measurements during photon irradiation, between LINAC pulses, to measure delayed gamma rays from short-lived fission products. Photofission measurement with a LaBr3 scintillation detector can help determining, in a Bayesian frame, the mass and position of actinides in radioactive waste packages. In addition to counting statistics, the model takes into account photofission yields and waste matrix density uncertainties. The model performances are numerically illustrated in the case of a 235 g238U lump placed inside a 50 cm radius drum homogeneously filled with 2.35 g/cm3 density concrete. In the most unfavourable angular position, with 32 drum rotations, when the 238U radial position is 10% and 90% of the waste drum radius, the expected relative uncertainty on the 238U mass is respectively 32% and 45%. The 238U angular position uncertainty is 10 mrad and its radial position can be determined with an expected uncertainty of 2 cm and 3 cm, when the 238U is respectively placed at 10% and 90% of the waste drum radius.
To investigate the possibility to characterize radioactive wastes using available information and measurements in a coherent frame, a Bayesian formalism is built to couple gamma-ray spectrometry and ...tomographic scans. The gamma ray spectrometry is performed with four High Purity Germanium (HPGe) detectors placed around the waste and scanning 2 cm thick slices of a radioactive waste drum in a Gamma Scanning mode. The tomography provides the density of the drum matrix and identifies heterogeneities. The unknown chemical composition of the matrix and heterogeneities is handled with two parameters that represent the mass fractions of carbon and hydrogen of a fictive {C; Sn; H} material that would show a gamma ray attenuation behavior similar to that of the true material. The approach is tested by simulating the measurement of 239Pu gamma rays produced by a PuO2 sphere placed in a drum slice in which two distinctive zones have been identified by tomography, in presence of 137Cs background. With the studied simulated examples, depending on the sphere position, the use of the density prior information allows decreasing the plutonium mass uncertainty by up to a factor 3 compared to the case when the density prior is not used. In addition, the use of the density prior information provides a posterior plutonium mass distribution having a significant number of event related to the true plutonium mass, which is not the case when the density prior information is not used.
MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using ...the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way.
Program title: MODAR
Catalogue identifier: AEGA_v1_1
Program summary URL:
http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_1.html
Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland
Licensing provisions: Standard CPC licence,
http://cpc.cs.qub.ac.uk/licence/licence.html
No. of lines in distributed program, including test data, etc.: 150 927
No. of bytes in distributed program, including test data, etc.: 4 981 633
Distribution format: tar.gz
Programming language: C++
Computer: Most Unix workstations and PCs
Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP.
RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two dimensional
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bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself.
Classification: 17.6
Catalogue identifier of previous version: AEGA_v1_0
Journal reference of previous version: Comput. Phys. Comm. 181 (2010) 1161
External routines: ROOT version 5.24.00 (
http://root.cern.ch/drupal/)
Does the new version supersede the previous version?: Yes
Nature of problem: The output of a MCNP simulation is an ascii file. The data processing is usually performed by copying and pasting the relevant parts of the ascii file into Microsoft Excel. Such an approach is satisfactory when the quantity of data is small but is not efficient when the size of the simulated data is large, for example when time-energy correlations are studied in detail such as in problems involving the associated particle technique. In addition, since the finite time resolution of the simulated detector cannot be modeled with MCNP, systems in which time-energy correlation is crucial cannot be described in a satisfactory way. Finally, realistic particle energy deposit in detectors is calculated with MCNP in a two step process involving type-5 then type-8 tallies. In the first step, the photon flux energy spectrum associated to a time region is selected and serves as a source energy distribution for the second step. Thus, several files must be manipulated before getting the result, which can be time consuming if one needs to study several time regions or different detectors performances. In the same way, modeling counting statistics obtained in a limited acquisition time requires several steps and can also be time consuming.
Solution method: In order to overcome the previous limitations, the MODAR C++ code has been written to make use of CERN's ROOT data analysis software. MCNP output data are read from the MCNP output file with dedicated routines. Two dimensional histograms are filled and can be handled efficiently within the ROOT framework. To keep a user friendly analysis tool, all processing and data display can be done by means of ROOT Graphical User Interface. Specific routines have been written to include detectors finite time resolution and energy response function as well as counting statistics in a straightforward way.
Reasons for new version: For applications involving the Associate Particle Technique, a large number of gamma rays are produced by the fast neutrons interactions. To study the energy spectra, it is useful to identify the gamma-ray energy peaks in a straightforward way. Therefore, the possibility to show gamma rays corresponding to specific reactions has been added in MODAR.
Summary of revisions: It is possible to use a gamma ray database to better identify in the energy spectra gamma ray peaks with their first and second escapes. Histograms can be scaled by the number of source particle to evaluate the number of counts that is expected without statistical uncertainties.
Additional comments: The possibility of adding tallies has also been incorporated in MODAR in order to describe systems in which the signal from several detectors can be summed. Moreover, MODAR can be adapted to handle other problems involving two dimensional data.
Running time: The CPU time needed to smear a two dimensional histogram depends on the size of the histogram. In the presented example, the time-energy smearing of one of the
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two dimensional histograms takes 3 minutes with a DELL computer equipped with INTEL Core 2.
In neutron Time-of-Flight (TOF) measurements performed with fast organic scintillation detectors, both pulse arrival time and amplitude are relevant. Monte Carlo simulation can be used to calculate ...the time–energy dependant neutron flux at the detector position. To convert the flux into a pulse height spectrum, one must calculate the detector response function for mono-energetic neutrons. MCNP can be used to design TOF systems, but standard MCNP versions cannot reliably calculate the energy deposited by fast neutrons in the detector since multiple scattering effects must be taken into account in an analog way, the individual recoil particles energy deposit being summed with the appropriate scintillation efficiency. In this paper, the energy response function of 2″×2″ and 5″×5″ liquid scintillation BC-501A (Bicron) detectors to fast neutrons ranging from 20keV to 5.0MeV is computed with GEANT4 to be coupled with MCNPX through the “MCNP Output Data Analysis” software developed under ROOT (Carasco, 2010).
► GEANT4 has been used to model organic scintillators response to neutrons up to 5MeV. ► The response of 2″×2″ and 5″×5″ BC501A detectors has been parameterized with simple functions. ► Parameterization will allow the modeling of neutron Time of Flight measurements with MCNP using tools based on CERN's ROOT.
In the frame of the effective Container inspection at BORDer control points (C-BORD) project H2020 program of the European Union (EU), a Rapidly Relocatable Tagged Neutron Inspection System (RRTNIS) ...has been developed for a nonintrusive inspection of cargo containers, aiming at explosives and other illicit goods detection. Twenty large-volume NaI detectors are used to determine the elements composing inspected materials from their specific gamma-ray spectra signatures induced by fast neutrons. The RRTNIS inspection is focused on a specific suspect area selected by X-ray radiography. An unfolding algorithm decomposes the energy spectrum of this suspect area on a database of pure element gamma signatures. A classification is then performed between inorganic materials, such as metals, ceramics, or chemicals, and organic materials like wood, fabrics, or plastic goods. Concerning organic materials, the obtained elemental proportions of carbon, nitrogen, and oxygen allow discriminating explosives from illicit drugs and benign substances. This article reports on the final laboratory tests performed at Commissariat à <inline-formula> <tex-math notation="LaTeX">\text{I}^\prime </tex-math></inline-formula>Énergie Atomique et aux Énergies Alternatives (CEA) Saclay, France, to assess the RRTNIS detection performances before further demonstration tests in a real seaport environment. Simulants of explosives and illicit drugs have been hidden at different depths inside iron or wood cargo materials, which are representative of the different neutron and gamma attenuation properties encountered in real cargo containers. Hundreds of experiments have been performed, showing that a few kilograms of explosives or narcotics can be detected by the RRTNIS in 10-min inspections.
The measurement of delayed gamma rays following neutron-induced fission is simulated with MCNP 6.1 to investigate the feasibility of fissile material detection in long-lived, medium activity ...radioactive waste in 870 L drums. The signal from homogeneously distributed fissile material in the drum is several hundred counts in the main delayed gamma peaks of interest. In a peripheral position or in the drum center, the signal is however too small to allow for a reliable measurement.
This paper reports on a numerical feasibility study of a passive neutron coincidence counting system for radioactive waste drums with plastic scintillators. The motivation is to replace 3 He gas ...counters generally used for this type of measurement. Indeed, plastic scintillators present several advantages for the measurement of neutron coincidences such as a good efficiency for detecting fast neutrons, short detection time, and low cost comparatively to 3 He. However, unlike 3 He counters, their high sensitivity to gamma rays and cross talk constitutes a drawback as parasite random and true coincidences are detected together with the useful signal of plutonium. Simulations are performed using the Monte Carlo transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. Performances of the coincidence counting system are studied for the case of a vitrified waste drum containing Pu and 241 Am, focusing particularly on multiplicity 1 and 2, i.e., 2 or 3 pulses recorded in a short time gate in different detectors. Cross talk induced by neutrons and gamma rays has been characterized in terms of time and distance between detectors, and strategies to limit this phenomenon are reported, consisting of ignoring neighboring detectors signal. A significant improvement of the Pu to 241 Am ratio for multiplicity 2 coincidences has thus been obtained, at the expense of counting statistics. Alternative case studies with organic and metallic matrixes of technological wastes are also reported, for which the part of useful signal of plutonium is significantly higher, showing the feasibility of the measurement method.
As part of its R&xD activities in the fields of radioactivewaste drum storage and homeland security, the NuclearMeasurement Laboratory of CEA Cadarache has started studiesrelated to the detection of ...induced delayed fission gamma rays asa signature of U/Pu presence either in radioactive wastes or incargo containers and luggage. The study described in the presentpaper explores the feasibility of detecting fission delayed gammarays of nuclear materials interrogated by a pulsed neutrongenerator. For this purpose, Monte Carlo simulations have beenperformed with ACT, the MNCP6 Activation Control Card.Simulated results have been compared with experimental data tovalidate the numerical model. Samples of uranium andplutonium have been irradiated for 2 hours with a pulsed D-Tneutron generator delivering 14 MeV neutrons with an averageemission of 8.10
7
n/s, which are thermalised in a graphite cellcalled REGAIN. At the end of irradiation, activated nuclearmaterials were placed in a low-background, high-resolutiongamma spectroscopy station in order to detect delayed gammarays emitted by fission products. Anomalies have been observedin the calculated time decay curve of fission delayed gamma rayswith MCNP6 ACT card, but the time behavior is correct for non-fission activated materials like aluminum or copper. On the otherhand, the number of counts recorded in the main simulatedgamma ray lines from activated nuclear material fission productsis consistent with the experimental results, thus validating thesimulation scheme in view of further studies on thecharacterization of radioactive waste drums or special nuclearmaterial detection in cargo containers.
AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration measurement by means of gamma-ray logging. The ...determination of uranium concentration in boreholes is performed with the Natural Gamma Ray Sonde (NGRS) based on a NaI(Tl) scintillation detector. The total gamma count rate is converted into uranium concentration using a calibration coefficient measured in concrete blocks with known uranium concentration in the AREVA Mines calibration facility located in Bessines, France. Until now, to take into account gamma attenuation in a variety of boreholes diameters, tubing materials, diameters and thicknesses, filling fluid densities, and compositions, a semiempirical formula was used to correct the calibration coefficient measured in Bessines facility. In this paper, we propose to use Monte Carlo simulations to improve gamma attenuation corrections. To this purpose, the NGRS probe and the calibration measurements in the standard concrete blocks have been modeled with Monte Carlo N-Particles (MCNP) computer code. The calibration coefficient determined by simulation 5.3 <inline-formula> <tex-math notation="LaTeX">\text{s}^{-1}\cdot \text {ppm}_{U}^{-1} </tex-math></inline-formula> with 10% accuracy is in good agreement with the one measured in Bessines (and for which no uncertainty was provided), 5.2 <inline-formula> <tex-math notation="LaTeX">\text{s}^{-1}\cdot \text {ppm}_{U}^{-1} </tex-math></inline-formula>. The calculations indicate that the concrete blocks used for measuring the calibration coefficients measured in Bessines are underestimated by about 10%. Based on the validated MCNP model, several parametric studies have been performed. For instance, the rock density and chemical composition proved to have a limited impact on the calibration coefficient. However, gamma self-absorption in uranium leads to a nonlinear relationship between count rate and uranium concentration beyond approximately 1% of uranium weight fraction, the underestimation of the uranium content reaching more than a factor 2.5 for a 50% uranium weight fraction. Parametric studies have also been performed with different tubing materials, diameters, and thicknesses, as well as different borehole filling fluids representative of real measurement conditions, in view to validate gamma attenuation corrections based on the semiempirical formula. In addition, a multilinear analysis approach has been tested to further improve accuracy on uranium concentration determination, leading to only a few percent uncertainties on a large range of configurations.
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